Post-irradiation Examination Results Research Articles

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (∼95 K/s) and peak surface temperatures (∼2527 K). These surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

Post-irradiation examination of U-7Mo/Mg and U-10Mo/Mg dispersion fuels irradiated in the NRU reactor

As part of a global initiative to reduce the usage of highly enriched uranium (HEU) fuel for research and test reactors, Al-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements with a low-enriched uranium (U-235 enrichment of 19.75 wt%) loading of 4.5 g U/cm3 were fabricated at the Canadian Nuclear Laboratories (CNL). The mini-elements were irradiated in the National Research Universal (NRU) reactor at maximum linear power ratings of 100 kW/m up to 80 at% U-235 burnup. Hot cell macroscopic examination of the discharged fuel elements at 30, 60 and 80 at% U-235 burnup indicated that they were intact without breakaway swelling or fuel cladding failure. The results of interim post-irradiation examination (PIE) that was conducted on mini-elements discharged at 30 at% U-235 burnup are reported in this paper. A follow-up paper will present the PIE results of the mini-elements that achieved burnups of 60 and 80 at% U-235. In the present study, the microstructure and phase composition of the irradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were investigated using optical microscopy and neutron diffraction analysis. The weight fraction of different crystalline phases in U-7Mo/Mg and U-10Mo/Mg fuel cores at low burnup was determined through Rietveld refinement. This paper includes a description of the irradiation tests along with preliminary PIE results from the irradiated U-Mo/Mg mini-elements and describes the results of optical microscopy characterization and neutron diffraction analysis.

Microstructure investigations of temperature effect on Al-UMo diffusion couples irradiated by swift Xe ions

Post-irradiation examination (PIE) results of microstructure in irradiated Al-UMo diffusion couples are reported. These diffusion couples were irradiated by 60 MeV Xe ions at four different temperatures up to approximately 5.5 × 1017 ions/cm2 peak ion fluence. An Al-UMo interaction layer (IL) was found to form at all four investigated temperatures. The IL is homogeneously amorphous when formed up to 150 °C. At 215 °C, the (U,Mo)Al3 nanocrystalline precipitates form within the amorphous IL matrix. Kirkendall voids with prominent temperature dependence were observed in both the Al layer and its interface with the IL. On the other hand, Xe bubbles were found to form in UMo and Al-UMo IL with different morphology. These microstructure features were quantitatively measured and are reported to provide valuable references for understanding the irradiation behavior of UMo/Al dispersion fuel.

FINITE ELEMENT ANALYSIS OF AN EMPTY 37-ELEMENT CANDU® FUEL BUNDLE TO STUDY THE EFFECTS OF PRESSURE TUBE CREEP

CANDU fuel bundles experience plastic deformations over time, and the horizontal configuration of the bundle in a crept pressure tube (PT) causes coolant to bypass the sagged lower half of the bundle. Bundle segments where the flow is limited may become more susceptible to dryout due to reactor aging. A finite element model of a 37-element fuel bundle was constructed using the commercial finite element software ANSYS to study the mechanical deformation behaviour of the bundle to maintain a coolable geometry. The main focus was on the contact between the fuel elements and between the fuel elements and PT. The complexity of the model due to all the contact pairs necessitated the use of high-powered computing hardware. Contact was demonstrated between the appendages, and sensitivity of the deformation to different boundary conditions (BC) was investigated. In particular, the radial position where the elements were welded to the endplate significantly impacted the magnitude of the element bowing. Expanding the PT up to 8% diametral creep demonstrated the proper functioning of the spacer pads (SP) and bearing pads in preventing sheath-to-sheath contact at the midplane and sheath-to-PT contact. However, the quarter plane was deemed to be the critical region due to the lack of SPs preventing excessive element bowing. This work has successfully illustrated the deformation of a CANDU fuel bundle, with contact, and its similarity with the bow profiles when compared with post-irradiation examination results and bundle heat-up tests.

Microstructure evolution in U-10Zr alloy irradiated by swift Xe ions at 700 ∘C

The post-irradiation examination (PIE) results of U-10Zr binary metallic fuel that was irradiated by 84 MeV Xe ions at 700 ∘C are reported. The U-10Zr specimen was synthesized by homogenizing elemental U and Zr powders through arc-melting. The as-fabricated U-10Zr was dominated by the sub-micron acicular α+δ phase. After being irradiated by 84 MeV Xe ions at 700 ∘C up to 2.19 × 1017 ions/cm2 fluence, the U-10Zr was speculated to evolve into micro-scale γ1+γ2 phases as predicted by the U-Zr phase diagram. During post-irradiation cooling, the U-rich γ1 phase was mainly transformed to the α phase, while the Zr-rich γ2 phase was partially retained along with formation of hexagonal structure phase (δ or ω). Xe bubbles formed in the γ1 and γ2 phases have salient different morphologies, implying faster Xe diffusion in the γ1 phase than the γ2 phase. The quantitative bubble morphology data collected in this study can provide valuable references for the development and validation of fuel performance models for this fuel.

Non-Destructive post-irradiation examination results of the first modern fueled experiments in TREAT

A series of experiments were conducted in the Transient Reactor Test (TREAT) Facility as part of commissioning for nuclear fuel safety research to be continued over the next several decades. TREAT was first constructed in the late 1950s to support research on nuclear fuel specimens under extreme nuclear-heated conditions. Following a decades-long hiatus, reactor operations were resumed at TREAT in 2017 to support the reemerging field of fuel-safety research. The first fueled tests were performed on fresh light-water reactor (LWR)-type sub length specimens (4.9% enriched UO2 in zirconium-alloy cladding) in inert gas capsules with the primary objective to demonstrate new irradiation capabilities, including calculation of energy-coupling factor, and to provide a baseline reference for interpretation of future experiments on accident tolerant fuel designs and high burnup standard fuels. In parallel, several new nondestructive post-irradiation experimental systems and processes have been developed and were successfully demonstrated using these samples. Five capsules were irradiated with energy injections and peak cladding temperatures of ~212 and ~1312 J/gUO2 and ~561–~2113 °C, respectively. These irradiations resulted in various degrees of fuel damage that were characterized using advanced non-destructive examinations. The post-irradiation examinations highlighted that, in the most-energetic transient, zirconium breakout, also referred to as ‘candling,’ was observed as gross cladding melting and relocation. Post-irradiation examinations also confirmed results inferred from in situ instrumentation during irradiation and are consistent with previous results from other studies.

Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Several different U based metallic nuclear fuel alloys and forms were irradiated as part of the Advanced Fuels Campaign (AFC) series of irradiation experiments. Results from postirradiation examinations of the AFC-3A and AFC-3B are presented here. In this irradiation, U alloyed with Mo, Zr, and Zr with Pd were investigated. Different geometric forms (solid, annular), smear densities and bonding media (sodium, helium) were also irradiated and evaluated. All fuel forms were irradiated in ferritic-martensitic HT-9 cladding. Examination of the irradiated fuel included neutron radiography, evaluation of the distribution of gamma-ray emitting radionuclides, dimensional inspection, fission gas release analysis, optical microscopy, and chemical analysis. No in-pile failure was observed with this fuel. Neutron radiography, gamma spectrometry, and dimensional inspections indicated the fuel largely performed acceptably. Optical microscopy revealed significant fuel-cladding interaction between the Mo based alloys and the cladding. The Zr based alloys generally performed better although some forms revealed behaviors that present specific fabrication challenges such as tighter fabrication tolerances that will need to be better controlled to match historical U-Zr fuel performance. The addition of Pd as an additive to prevent fuel cladding chemical interaction (FCCI) also appears to present some challenges to fuel performance.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compacts containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. In the case of krypton, all the coating layers are partly retentive and the available data did not allow the level of retention in individual layers to be determined, hence preventing derivation of any correction factors.

INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

Simulation of Pellet-Cladding Interaction with the PLEIADES Fuel Performance Software Environment

This paper focuses on the PLEIADES fuel performance software environment and its application to the modeling of pellet-cladding interaction (PCI). The PLEIADES platform has been under development for 10 yr; a unified software environment, including the multidimensional finite element solver CAST3M, has been used to develop eight computation schemes now under operation. Among the latter, the ALCYONE application is devoted to pressurized water reactor fuel rod behavior. This application provides a three-dimensional (3-D) model for a detailed analysis of fuel element behavior and enables validation through comparing simulation and postirradiation examination results (cladding residual diameter and ridges, dishing filling, pellet cracking, etc.). These last years the 3-D computation scheme of the ALCYONE application has been enriched with a complete set of physical models to take into account thermomechanical and chemical-physical behavior of the fuel element under irradiation. These models have been validated through the ALCYONE application on a large experimental database composed of approximately 400 study cases. The strong point of the ALCYONE application concerns the local approach of stress-corrosion-cracking rupture under PCI, which can be computed with the 3-D finite element solver.Further developments for PCI modeling in the PLEIADES platform are devoted to a new mesh refinement method for assessing stress-and-strain concentration (multigrid technique) and a new component for assessing fission product chemical recombination.

Simple formula to evaluate helium production amount in fast reactor MA-containing MOX fuel and its accuracy

A simple formula is developed for the evaluation of the helium production amount in fast reactor minor actinide (MA) containing uranium–plutonium mixed oxide (MOX) fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, it is found that the formula evaluates the helium production amount with the difference of less than about 10% from the detailed calculation and the PIE results, when the MA isotope content is less than 5 wt.%. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

Physical Properties and Irradiation Behavior Analysis of Np- and Am-Bearing MOX Fuels

The physical properties and irradiation behavior of Np- and Am-bearing MOX were evaluated for the development of advanced fast reactor fuels. The physical properties, lattice parameter, melting temperature, thermal conductivity, and oxygen potential of the fuels were described as functions of Np content, Am content, and oxygen-to-metal (O/M) ratio, and the effect of a few percent Np and Am addition into MOX on those properties was found to be negligibly small. The irradiation tests of fuel pellets having O/M ratios of 1.98 or 1.96 were carried out at a high linear heat rate of about 430 W/cm for 24 h. The redistributions of actinide element and oxygen were analyzed by using the physical properties for Np- and Am-bearing MOX, and the maximum temperatures of the pellets were estimated. The maximum temperatures of the pellets of O/M = 1:96 and 1.98 were estimated to be 2,701 and 2,510 K, respectively. The analysis result showed that the higher maximum temperature in the low-O/M pellet was caused by the low thermal conductivity in the outer region of the pellets, and the maximum temperatures of the pellets did not exceed their melting temperatures. In the results of postirradiation examination, no trace of melting was observed.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20–40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U–Pu–10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U–15Pu–6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.

PERFORMANCE EVALUATION OF U-Mo/Al DISPERSION FUEL BY CONSIDERING A FUEL-MATRIX INTERACTION

Because the interaction layers that form between U-Mo particles and the Al matrix degrade the thermal properties of U-Mo/Al dispersion fuel, an investigation was undertaken of the undesirable feedback effect between an interaction layer growth and a centerline temperature increase for dispersion fuel. The radial temperature distribution due to interaction layer growth during irradiation was calculated iteratively in relation to changes in the volume fractions, the thermal conductivities of the constituents, and the oxide thickness with the burnup. The interaction layer growth, which is estimated on the basis of the temperature calculations, showed a reasonable agreement with the post-irradiation examination results of the U-Mo/Al dispersion fuel rods irradiated at the HANARO reactor. The U-Mo particle size was found to be a dominant factor that determined the fuel temperature during irradiation. Dispersion fuel with larger U-Mo particles revealed lower levels of both the interaction layer formation and the fuel temperature increase. The results confirm that the use of large U-Mo particles appears to be an effective way of mitigating the thermal degradation of U-Mo/Al dispersion fuel.

Irradiation Testing of High-Power-Density Vibropacked Annular Fuel

This paper describes an irradiation test of high-power-density internally and externally cooled annular fuel samples in the 5-MW Massachusetts Institute of Technology (MIT) research reactor MITR-II. The design of the irradiation facility is briefly reviewed, with an emphasis on the thermal-hydraulic behavior of the irradiation capsules. The irradiation test is described, including the thermal history of the two irradiated samples. A discussion of the observed asymmetrical temperature profiles is provided. Results of preliminary postirradiation examination consisting of collimated gamma scans of the irradiation capsules to confirm burnup estimates and estimate fission gas release (FGR) are also presented. It is concluded that the vibropacked fuel samples’ FGR is below 1%, and that is within the predictable range by a specially equipped FRAPCON model.

Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance

Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally has displayed gas release values during irradiation three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

Phase equilibria in the BaUO3-BaZrO3-BaMoO3 system

The phase equilibria in the pseudo-ternary BaUO3-BaZrO3-BaMoO3 system were studied to understand the thermochemical properties of the perovskite type gray oxide phase in high burnup MOX fuel. Thermodynamic equilibrium calculation for the system was performed by using a ChemSage program under the various oxygen potentials. Solid solutions existing in the system were treated by an ideal solution model. The present calculation results well agreed with the previous reported post irradiation examination results, showing that BaMoO3 was scarcely included in the gray oxide phase.